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At least 1,027 records · Page 57

Remote In-bore Laser Cutting and Welding Tools for use in Future Nuclear Fusion Reactors - 18348

The maintenance, replacement and decommissioning of future nuclear fusion reactors will require quick and reliable cutting and joining of in-vessel pipework. Initial design studies for nuclear fusion reactors for power generation have estimated cutting and welding could account for up to 60% of the maintenance duration using conventional in-situ processing techniques (dry-mechanical cutting and arc welding) and that new methods are required to expedite the process. Additionally, the expected radioactivity and limited access at the cutting and welding sites mean that these processes cannot be done manually and robotic tools are required. To this end, remote in-bore laser cutting and welding tools have been developed for use in 90 mm internal diameter steel pipes. Here, we will present the designs of the remote in-bore laser cutting and welding tools, and how the tools fit within the overall remote maintenance strategy of a nuclear fusion reactor. Initial high-power laser trials were performed with prototype tools which successfully demonstrated their laser processing, thermal management and dust protection functionality. Although the tools described herein have been developed specifically for use within the fusion reactor environment, however the tool design and technology demonstrated here is readily transferable to many remote applications in challenging environments such as fission reactor maintenance, nuclear decommissioning and other in-accessible pipework. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Preliminary results of AGC-4 irradiation in the advanced test reactor and design of AGC-5 - 18469

The U.S. Department of Energy Advanced Reactor Technologies Program will irradiate up to six nuclear graphite creep experiments in Idaho National Laboratory Advanced Test Reactor (ATR). The graphite experiments are being irradiated over an approximate 8-year period to support development of a graphite irradiation performance database on new nuclear grade graphites now available for use in high-temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the very high-temperature gas reactor and other future gas reactors. Each of the experiments consist of a single capsule that contains six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load and the other half of specimens not subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of the diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with online temperature and compressive load monitoring and control. Also, samples are taken of sweep gas effluent to measure any oxidation or off-gassing that may occur during initial startup of the experiment. The first experiment Advanced Graphite Capsule (AGCJ-I started irradiation in September 2009 and completed in January 2011; the second experiment AGC-2 started irradiation in April 2011 and completed in May 2012; the third experiment AGC-3 started irradiation in late November 2012 and completed in April 2014; and AGC-4 is currently being irradiated in ATR. This paper will briefly discuss preliminary irradiation results from the AGC-4 experiment and design of AGC-3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Preliminary analysis of effect of Vanadium Self-Powered Neutron Detectors on AP1000 reactor core physical parameters - 360

For the AP1000 reactor, the Vanadium Self-Powered Neutron Detectors are installed in the instrumentation thimbles of the fuel assembly. The (n,? ) reaction of Vanadium SPNDs will impact on the neutron spectrum of the location, and finally impact on the reactor core physical parameters. But, this effect is not considered for the AP1000 reactor Nuclear Design. So, it is necessary to analyze the effect of the presence of Vanadium SPNDs on the AP1000 reactor core physical parameters. CASMO5 is used to calculate physical parameters of four typical assemblies of AP1000 reactor, and SIMULATE5 is used to calculate physical parameters of the initial cycle. The results show that the maximum absolute deviations of K-infinite, pin power, thermal flux for the assemblies are 181 pcm, 6.3% and 20.1%, separately. The maximum absolute deviations of critical boron concentration, heat flux hot channel factor, nuclear enthalpy rise hot channel factor, axial offset, quadrant power tilt ratio and axial power distribution of the initial cycle are 5 ppm, 0.7%, 0.5%, 0.1%,0.09% and 0.19%,separately. So, the presence of Vanadium SPNDs has a significant impact on the power distribution of pin cells around the instrumentation thimble and the thermal flux in the location of instrumentation thimble, but doesn't affect core physical parameters of the initial cycle very much. As is shown, it is very important to accurately consider the effect of presence of Vanadium SPNDs when calculating the Vanadium SPNDs current response. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Derivation of the Point-Kinetics Parameters for the Coupled Reactors

The design of coupled reactors, which provide two distinct spectral zones (fast and thermal), has recently drawn attention because it can provide configurations that could not be met in current thermal neutron test reactors and thus allows versatile irradiation tasks. A coupled reactor core is characterized by the neutron multiplicity of each zone as well as by the coupling between the two zones. It is desired to evaluate the integral point kinetics parameters that characterize the neutron physics of the coupled system. This summary provides a derivation of the equations for evaluating these point kinetics parameters. Following the strong and the weak form of the multigroup diffusion equation, the weak forms for the partial forward and adjoint solutions, each for a particular active region of the reactor core, are derived. These partial solutions are then used to collapse the weak form into the coupled-reactor point kinetics equations with the approximation of equilibrium. It is noted that these parameters are not meant for transient analysis because of the equilibrium approximation. They are rather used for characterizing the coupled system. The derived reactivity equations are then compared with those in Ref. 1. The equations are implemented in Rattlesnake, the INL MOOSE-based radiation transport application for multi-physics simulations. Numerical results are presented for a simple two-group problem. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of a Research Reactor Protocol for Neutron Multiplication Measurements

Establishing a protocol for measurements on research reactors is the next step in advanced subcritical neutron multiplication inference measurements. Such measurements can help identify deficiencies and quantify uncertainties in nuclear data and validate predictive Monte Carlo radiation transport simulation capabilities related to subcritical neutron multiplication inference techniques. This work expands on previous benchmark-quality efforts performed at the National Criticality Experiments Research Center (NCERC) located in Nevada, including subcritical experiments with a 4.5-kg {alpha}-phase plutonium sphere surrounded by copper, tungsten and nickel. Evaluations of the nickel and tungsten measurements have both been accepted into the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The nickel benchmark was the first ICSBEP-accepted evaluation of measurements using the Feynman Variance-to-Mean method based on the Hage-Cifarelli formulism, and was the culmination of many years of subcritical experiment research. To help establish a protocol for research reactor-based subcritical measurements, a series of Critical and Subcritical {Phi}-Power Experiments at Rensselaer (CaSPER) have been designed. This neutron multiplication measurement series will be executed at the Rensselaer Polytechnic Institute Walthousen Reactor Critical Facility (RPI-RCF) in upstate New York. The RPI reactor facility can provide benchmark-quality integral experimental configurations where different reactivity states can be achieved by varying the control rod height/water height in the reactor. The diversity of achievable configurations are also unique in contrast to previous subcritical benchmark measurements in that they are spatially complex, involve different materials (fuel, moderator), and are characterized by system-specific neutron cross section sensitivities (various energy ranges and neutron reaction contributions). It is worth noting that previous proof-of-concept subcritical measurements have been performed on thermal systems at NCERC, at a variety of reactivity states by separation distance, yet it was not possible to know the reactivity of various configurations accurately for benchmarking (there are no control rod worth curves as there are no control rods). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Phenomena Identification and Ranking Table (PIRT) for Fluoride High-Temperature Reactor (FHR) Neutronics

The deployment of Fluoride (salt-cooled) High Temperature Reactors (FHRs) has many potential benefits such as improved safety, improved economics, and a diversification of the energy portfolio because of its capability to supply industrial process heat. However, significant challenges remain before this class of reactors can be deployed, mostly related to its technology readiness. A recent study identified several areas in need of further development in order to advance the technology. These include the verification and validation of neutronics and thermal hydraulics modeling and simulation tools in support of design, safety analysis and licensing. The focus of this paper is to summarize the issues for verification and validation of neutronics tools. Current tools for neutronics analysis of LWRs are inadequate for the simulation of salt cooled reactors. The methodologies developed for the high temperature gas cooled reactors (HTGRs) have some relevance for FHRs because of the similarities between the two reactors such as graphite block structure, encapsulated fuel particles, and high operating temperatures. However, they are not sufficient for the accuracy desired of a code for licensing as they have a number of important differences neutronically, such as the use of liquid salt containing beryllium as a coolant and their unique fuel configuration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Comparison of Proposed HANARO Fuel Assemblies in the WWR-SM Reactor

Fuel supply stability is crucial to all reactors. The WWRSM research reactor operated by the Institute of Nuclear Physics at the Academy of Sciences in Uzbekistan currently uses Russian-supplied IRT-4M fuel. The goal of this Oregon State/PNNL research collaboration is to investigate the potential use of the Korean-made High-flux Advanced Neutron Application Reactor (HANARO) fuel rods in the WWR-SM. Many research reactors use unique fuels made by single suppliers, like the WWR-SM. The WWR-SM is a pool-type, high power research reactor using IRT-4M fuel assemblies which are sold and manufactured by a single fuel supplier. Use of HANARO fuel rods has the potential to improve fuel supply stability and associated costs while maintaining high flux levels needed for continuing operations. In the course of the investigation to design an alternate fuel assembly for the WWR-SM, multiple designs have been evaluated. In particular, two assemblies using standard HANARO fuel rods were considered: one with the optimal fuel-to-moderator ratio in the fueled region of the assembly and one with an under-moderated fuel region. Each of these assemblies has had its neutronic performance characterized with MCNP6 system reactivity and flux distribution calculations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Preliminary Analysis of Damage to MPFDs Caused by Reactor Pulses

The need for next-generation, accident-tolerant fuels has generated significant interest in the restart of domestic fuels testing at the Transient Reactor Test (TREAT) facility. The TREAT reactor is designed to produce large, transient fluxes that replicate the effects of a wide variety of accident conditions on an irradiated fuel sample. Peak powers generated during short pulses approach 20 GW with a neutron flux of approximately 10{sup 17} n cm{sup -2} s{sup -1}. As part of the restart effort, new vehicle designs are being explored for sample irradiation, including the Static Environment Rodlet Transient Test Apparatus (SERTTA) and variants. As part of the SERTTA designs, micro-pocket fission detectors (MPFDs) are being explored for localized flux measurements before, during, and after transient operations. These detectors are also being considered as part of future, core-wide instrumentation for the TREAT reactor. However, the high fluxes associated with the pulses in the TREAT core present certain thermodynamic concerns given their intensity and the amount of energy they will release within the MPFDs within such a short amount of time. Assuming each fission that occurs in a MPFD results in a fission fragment depositing 85 MeV of energy in the substrate on average and each MPFD has 0.5 μg of natural uranium deposited on it, a 100 ms long pulse in the TREAT reactor at 1017 n cm{sup -2} s{sup -1} flux will produce approximately 4.6 x 10{sup 6} fissions in an MPFD. This pulse deposits 6.3 x 10{sup -5} J in the substrate over a 100 ms duration. That amount of energy is sufficient to melt roughly 33% of the Pt under the UO{sub 2} deposition or a circular spot with a diameter of approximately 860 μm. Of course, the above estimate makes a number of simplifying assumptions and ignores heat transfer, but does serve to illustrate the point that this rapid heating effect could be detrimental to the integrity and life of these in-core MPFDs. To study how these layers are affected by large transient neutron fluxes, five samples were prepared, pre-analyzed, and then subjected to a pulse in the central thimble of the Kansas State University TRIGA Mark II reactor to a peak power of 1.58 GW. After being allowed to decay to safer radiation levels, the sample substrates and coatings were inspected using scanning electron microscopy (SEM) to identify any changes to their surface morphology. Indeed, certain changes were found and are discussed in greater detail in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Monte Carlo/CFD Coupling for Accurate Modeling of the Delayed Neutron Precursors and Compressibility Effects in Molten Salt Reactors

Multiphysics reactor analysis has become one of the leading activities of most nuclear engineering research groups. Coupled neutronics and thermal/hydraulics (T/H) simulations have traditionally been performed by adopting deterministic transport tools and reactor system T/H codes. The main reason for avoiding higher fidelity numerical methods like continuous-energy Monte Carlo analysis has been the high computational cost of coupled, full-scale simulations. Nowadays, high fidelity multiphysics simulations represent a hot research topic, with several applications involving advanced nuclear systems. Liquid fuel Molten Salt Reactors (MSRs) feature peculiar physical phenomena that are not present in solid fuel reactors. Some of these phenomena are not correctly resolved by legacy reactor physics tools, and require specific treatments. This work focuses in particular on the analysis of the effect of the delayed neutron precursors drift and the impact of liquid fluid thermal expansion and compressibility in postulated criticality excursion transients in fast-spectrum MSRs. A new modeling tool based on a Monte Carlo approach for neutronics (Serpent code) coupled to a Computational Fluid Dynamic (CFD) analysis (OpenFOAM) was developed to obtain accurate modeling and simulation of these phenomena. The development of the Serpent Monte Carlo code [1] at VTT in the past few years has made this code particularly attractive for performing multiphysics studies. The availability of a flexible multiphysics interface, on-the-fly Doppler broadening and density-based collision rejection, and Delta-tracking, proved of paramount importance for the computationally efficient performance of coupled Monte Carlo/CFD calculations [2]. OpenFOAM is a C++ multiphysics toolkit featured by automatic matrix construction and solution capabilities for scalar and vector partial differential equations (PDEs), adopting several possible differencing and interpolation schemes, based on the Finite-Volume method. The ease of implementing the constitutive equations of different physical phenomena makes OpenFOAM particularly suitable for the modeling of complex coupled phenomena. In the past, several research activities proposed the adoption of both Serpent and OpenFOAM for the modeling and simulation of MSRs and the associated fuel cycles. In this work, these tools have been internally coupled for the steady-state and transient analysis of fast-spectrum systems, accounting for the effect of delayed neutron precursors transport and the salt thermal expansion and compressibility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Proceedings of the international conference on physics of reactors - Physor 2022

The theme for this year's conference is 'Making Virtual a Reality: Advancements in Reactor Physics to Leap Forward Reactor Operation and Deployment'. As we leap forward to the future of nuclear technology, micro-reactors, small modular reactors, advanced reactors, and high energy LWR fuel are at the forefront of the topics.The Physor 2022 will cover a broad range of topics in thirty standard and special sessions tracks, and ten panel sessions gathering more than 350 papers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologies (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Design of additively manufactured methanol conversion reactor for high throughput production

Additively manufactured methanol conversion reactor was designed and fabricated for high throughput fuel production while capturing carbon dioxide. Here, the reactor designed is part of large effort in integrating electrolyzer and catalytic reactor system. Processing in fabricating the steel, titanium, and nickel alloy reactors were optimized to improve the catalyst deposition process and reaction efficiency. Microstructure of the additively manufactured reactors were characterized. General design is utilized to improve the performance of the reactor while exploring the benefit of freeform in additive manufacturing.

08 HYDROGEN↗

Toward the implementation of self-powered, wireless, real-time reactor power sensing

The successful development of real-time, self-powered, wireless reactor power sensing technology would enable key operating parameters of every fuel assembly in a commercial reactor core to be continuously monitored, and potentially be utilized for applications such as closed-loop reactor control. This paper discusses the expansion of successful work on the application of Self Powered Neutron Detectors (SPND) for reactor power monitoring and the wireless transmission of these detector signals using advanced vacuum micro-electronic (VME) transmitter technology developed by Westinghouse in collaboration with the Penn State University Radiation Science and Engineering Center. Results demonstrate that this technology may be successfully applied via long-term testing in the Penn State TRIGA reactor. SPND are inherently slower than, for example, ion chambers due to their dependence on radioactive decay to generate a signal. Additional research developed delay compensation algorithms to accelerate the response of these detectors for use in near-real-time closed-loop reactor control. Initial results indicate that practical signal acceleration can be obtained digitally or with an equivalent analog circuit. Future research will apply an inverse model of the sensor’s balance differential equations to more-effectively accelerate the SPND signal. Use of inverse models will also prove to be adaptive in the sense that time-dependent burn-up of these detectors, due to neutron activation of the detector material, can be modeled and applied in the acceleration algorithm.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

High-burnup boiling water reactor steady-state operating conditions and fuel performance analysis

The primary operational costs for existing nuclear reactors are plant operation costs, maintenance costs, and fuel costs, all of which are influenced by the materials used and the design of the reactor core. Optimizing core design parameters—including burnup limits and enrichment levels—can lengthen cycles, reduce outages, reduce reload batch fractions and spent fuel storage requirements, and lower maintenance and operating expenses, thereby enhancing economic viability. Furthermore, developing higher-fidelity tools to simulate these parameters enables better identification of the available margin, improves overall plant safety, and improves the understanding a given plant’s responses to accident scenarios. Here, in the US, much of the research and development focus has traditionally been on pressurized water reactors (PWRs), but boiling water reactors (BWRs) comprise approximately one-third of the US reactor fleet. Modeling and simulation advances for BWRs and PWRs—particularly those achieved through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program—are crucial to the long-term viability of the light–water reactor industry. A key research area of the high burnup and increased enriched fuel initiative is focused on addressing issues related to postulated loss-of-coolant accident (LOCA) scenarios. NEAMS has dedicated significant effort to enhancing tools to better support BWRs. A current focus is showcasing the BWR framework for high-burnup LOCA analysis. This high-fidelity steady-state analysis is a first step toward demonstrating a best-estimate, pin-by-pin high-burnup BWR LOCA analysis to assess full-core cladding rupture behavior for a representative BWR. The objective of this effort is to provide a modeling capability that will help elucidate and provide a best-estimate evaluation for cladding rupture susceptibility in BWRs. This modeling capability could then be used to prevent and/or mitigate cladding ruptures in postulated accident scenarios without penalizing operational parameters. Additionally, the results of this work will help identify strategies for finding additional margins or potentially limiting cladding ruptures through core design optimizations to enable more efficient core designs.

Capps, Nathan [Oak Ridge National Laboratory (ORNL↗

Process intensification of biomass fast pyrolysis through autothermal operation of a fluidized bed reactor

Heat transfer is the bottleneck to fast pyrolysis of biomass. Although the enthalpy for pyrolysis of biomass is relatively small operation at temperatures around 500 °C constrains heat carrier selection to inert gases and granular media that can sustain only modest thermal fluxes in practical pyrolysis systems. With heat transfer controlling the rate of pyrolysis, reactor capacity only scales as the square of reactor diameter and does not benefit from economies of scale in building larger reactors. We have eliminated this heat transfer bottleneck by replacing it with partial oxidation of pyrolysis products to provide the enthalpy for pyrolysis in a fluidized bed reactor, a process that can be described as autothermal pyrolysis. Here, the oxygen-to-biomass equivalence ratio depends upon the kind of biomass being pyrolyzed and the level of parasitic heat losses from the reactor, but under conditions that simulate adiabatic operation, equivalence ratios are around 0.10, compared to 0.20 or higher for autothermal gasifiers. At this low equivalence ratio, there was no significant loss in bio-oil yield when operating the reactor autothermally (64.8 wt%) as compared to conventional pyrolysis (64.4 wt%). Carbon balances indicate that less valuable pyrolysis products (char and aqueous, bio-oil light ends) are consumed via partial oxidative reactions to provide the enthalpy for pyrolysis. While the carbon yields of char and bio-oil light ends decreased by 25.0% and 21.3%, respectively, the most valuable pyrolysis product (bio-oil heavy ends) only decreased 8.0%.

09 BIOMASS FUELS↗

Mass optimization of a supercritical CO 2 Brayton cycle with a direct cooled nuclear reactor for space surface power

A long life, reliable, and compact surface power system will be necessary to achieve future goals in space exploration. This application uniquely requires the system to be optimized with respect to mass because system mass directly drives the space launch cost associated with transporting the system to its desired location. In this study, a supercritical CO 2 Brayton cycle coupled to a direct-cooled nuclear reactor was designed and optimized for mass. Robust models were developed for each Brayton cycle component in order to model the cycle performance. The three most massive components of this cycle are the radiator-based heat rejection system (subsequently referred to as the “radiator”), recuperator, and reactor. Mass correlations for the recuperator and radiator were established through interactions with component experts from industry and national labs. A reactor model was developed to predict the minimum mass reactor that satisfies neutronic and thermal limitations for given cycle conditions. The system optimization explores tradeoffs between the reactor, recuperator, and radiator sizes in order to identify the least massive system that will satisfy the power (40 kWe) and life (10 yr) requirements. To explore the effects of turbine inlet temperature on system mass, three types of microtube and shell recuperator technologies were considered: baseline stainless steel (which is consistent with the industry partner’s current designs), stainless steel with non-heritage tube sizes (which requires further development of manufacturing techniques), and Inconel (which is not a current/legacy design). Both stainless steel designs have a temperature limit of 823 K, which limits the turbine inlet temperature to 900 K. The baseline stainless steel design results in a combined mass of 738 kg. The stainless steel design allowing for non-heritage tube sizes reduces the combined mass to 674 kg (a 9% improvement). The Inconel design leads to an optimal turbine inlet temperature of 1120 K and reduces the combined mass to 391 kg (a 47% improvement). Furthermore, the interesting conclusion from this study include: (1) radiator mass dominates the total mass, and this drives the cycle to relatively high heat rejection temperatures resulting in a compressor inlet state point that is not close to the vapor dome; thus, the typical advantages of an sCO 2 system are not realized, and a working fluid with a higher critical temperature may be more suitable, and (2) neutronic limitations cause the reactor size to be relatively unaffected by the power level.

42 ENGINEERING↗

Nanoparticle dynamics in the spatial afterglows of nonthermal plasma synthesis reactors

Nonthermal plasma flow tube reactors are industrially scalable systems for the production of nanocrystal (NC) based materials and coatings. One key advantage of nonthermal plasma synthesis is the ability to both synthesize NCs and deposit films in a single reactor, as at the reactor outlet, NCs can be inertially deposited onto a target substrate. The size and morphology of deposited particles can substantially influence the film structure and function. Though NCs are typically near-spherical and monodispersed as-produced in plasma synthesis reactors, NC charge and growth dynamics can be altered substantially when NCs are sampled out of the plasma and through the spatial afterglow region, affecting deposition. Experiments have demonstrated changes of NC size and charge in the spatial afterglow; however, these dynamics remain unexplored and unexplained via theory and simulation. To address this, we developed a constant number Monte Carlo (CNMC) simulation model to examine the mechanisms of NC decharging and growth in the spatial afterglow of plasma flow tube reactors. Collisions between NC and plasma species, diffusive deposition, and electron desorption from NCs are incorporated in the CNMC simulation. The simulation results are specifically compared with previous experiments on Si NCs synthesized from a low pressure Ar-SiH 4 nonthermal plasma reactor. Furthermore, the experiment-model comparison shows that CNMC models can be implemented which accurately model NC size distribution evolution in a spatial afterglow. Simultaneously, results show that improved collision models, energetic species diffusion models, and electron desorption models will be necessary to accurately depict NC dynamics in spatial afterglows.

42 ENGINEERING↗